Method of processing nuclear fuels

ABSTRACT

In the processing of irradiated nuclear fuels containing uranium and plutonium, the fuel is first converted to a plutoniumcontaining alkali-metal uranate, and the uranate is suspended in a molten salt or salt mixture and treated with gaseous hydrogen chloride and oxygen, whereupon the uranium dissolved in the melt is separated from the undissolved plutonium compounds.

United States Patent 1191 Avogadro et al. 1 1 Jan. 2, 1973 [54] METHODOF PROCESSING NUCLEAR 3,154,379 10/1964 Benedict ..23/3z5 FUELS3,160,470 l2/1964 Lambert .13/325 [75] Inventors: Alessandro Avogadro,Varese, Italy; FOREIGN PATENTS OR APPLICATIONS Joseph Wurm, Mol, Belgium663,943 11/1965 Belgium ..23/325 [73] Assignee: European Atomic EnergyCommuni- Kirehberg, LuXem- Primary Examiner-Carl D. Quarforth bourgAssistant Examiner-R. E. Schafer [22] Filed: sepL 8, 1969Attorney-Stevens, Davis, Miller & Mosher [21] Appl. No.: 856,171 57ABSTRACT In the processing of irradiated nuclear fuels containing [30]Forelgn Apphcatlon Pnomy Dam uranium and plutonium, the fuel is firstconverted to a Oct. 25, 1968 Netherlands ..68l530l Plutonium-containingalkali-metal uranate, and the uranate is suspended in a molten salt orsalt mixture [52] U.S. Cl ..423/5, 423/11 and treated with gaseous y ogechloride and [51] Int. Cl. ..G2lc 19/48 yg upon the u a ium dissolved inthe melt is [58] Field of Search ..23/325, 332 separated from theundissolved plutonium compounds. [56] References Cited UNITED STATESPATENTS 12/1961 Benedict ..23/325 1 Claim, N0 Drawings METHOD OFPROCESSING NUCLEAR FUELS This invention relates to a method ofprocessing uranium and plutonium containing irradiated nuclear fuels, inwhich the uranium from the fuel is converted into a plutonium-containingalkali-metal uranate.

A method of this type has been disclosed by British patent specificationNo. 1,108,042 according to which the nuclear fuel, which consistsprincipally of uranium oxide or in which the uranium has previously beenconverted to the oxide, is treated in a molten basic alkalimetalcompound in the presence of an oxidation agent, e.g. in molten NaOH or amixture thereof with LiOl-l, in the presence of air, oxygen or sodiumperoxide. From German patent specification No. 1,197,630 it is known tocarry out this process in molten nitrates, e.g. NaNO In these methods,the fuel forms as a pulverulent product which is insoluble in the meltand which consists principally of alkali-metal uranate, while thosefission products which are gaseous at the treatment temperature (400 to500 C.), e.g. krypton and xenon, escape and most of the fission productswhich are solid or liquid at said temperature, e.g. zirconium, niobium,cerium, barium and lanthanum and their compounds, dissolve in the melt.

Under the conditions described, plutonium forms insoluble compounds andon precipitation of the pulverulent alkali-metal uranate from the meltthe material obtained is largely free of fission products but stillcontains all the plutonium originally present. Generally, this plutoniumwill be more valuable than the used and impoverished uranium and it istherefore important to separate it from the uranium in an optimumquantitative yield. If this can be done simply at the site of thenuclear reactor, the high costs of transportation of irradiated fuelelements to processing plants can be obviated.

US. Pat. No. 3,01 1,865 discloses a process in which a mixture of theoxides of uranium and plutonium is suspended in a molten salt or saltmixture and is treated at elevated temperature (above 750 C.) with achlorination agent, namely C1 C1 plus HCl or with phosgene. The uraniumand plutonium dissolve in the melt. The separation of these elements isthen carried out by electrolysis or selective precipitation.Considerable disadvantages of this method are the high temperaturerequired and the corrosion problems entailed, and the fact that theuranium and plutonium separation is not complete.

It has now been found that these disadvantages can be obviated orreduced while in addition it is advantageously possible to useplutonium-containing alkali-metal uranate obtained in the mannerdescribed hereinbefore, if, as constitutes the characteristic featuresof the invention said uranate is suspended in a molten salt or saltmixture and is treated with gaseous hydrogen chloride and oxygen,whereupon the uranium dissolved in the melt is separated from theundissolved plutonium compounds.

Preferably, in the method according to the invention, in a first stage,the nuclear fuel is oxidized in a molten salt bath consisting primarilyof alkali-metal nitrates or peroxides. This oxidation, which can becarried out at a temperature below 450 C., yields a finely dividedproduct which is insoluble in the melt and which consists principally ofalkali-metal uranate and plutonium compounds. The oxidation is selectivewith respect to the fuel so that any cladding materials present, e.g.pieces of stainless steel or zirconium alloys, are not attacked. Thevolume of the fuel increases during this treatment so that it becomescompletely detached from the cladding. If this treatment is applied to afuel element which has been sawn open, e.g. a fuel rod clad with astainless steel can, the can from which the fuel has been eliminated canbe removed from the melt after some time while the alkali-metal uranateremains at the bottom of the crucible. This uranate is so finely dividedthat the subsequent chlorination can take place easily. Theplutonium-containing uranate is then separated from the melt, e.g. byfiltration over porous graphite and treated with hydrogen chloride andoxygen. To this end, it is first suspended in a molten salt or saltmixture, e.g'. tfie'eutectic mixture KCl and LiCI" or of MgCl and NaCl.During this chlorination, the uranate is selectively converted to uranylchloride, which dissolves in the melt, while the plutonium remains inthe form of insoluble compounds (oxide) and can be filtered off. Toobtain optimum separation, it is preferable to use more, by volume,oxygen than hydrogen chloride gas, i.e. a gas mixture containing morethan percent of oxygen by volume.

It is important that this selective chlorination should take place attemperatures of 450 C. and lower, so that it can be carried out in amelting crucible consisting, for example, of pyrographite withoutspecial corrosion problems. After filtration of the solid plutoniumcompounds, the uranyl chloride in the melt can be converted in knownmanner to uranium oxide by means of gaseous ammonia or the melt can bereacted with magnesium oxide, uranium oxide being precipitated and theresulting magnesium chloride remaining in solution in the melt.

The resulting plutonium oxide can be purified in known manner. Apartfrom the decontamination of the uranium and plutonium in the first stageas already described, in the second stage of the method the plutonium isalso separated from fission products such as the rare earths, chromiumand strontium which were still present after the first stage and whichdissolve in the melt in the form of their compounds.

Apart from materials such as pyrographite, the chlorination stage canalso be carried out using conventional ceramic materials, e.g. A1 0 orporcelain or ordinary graphite coated with a layer of silicon carbide.All the other treatments can be carried out by means of material whichis less resistant to corrosion, e.g. stainless steel or other nickelalloys. Thus all the filtration operations can be carried out by meansof filter plates consisting of metal or sintered ceramic material. Afterthe second stage of the process, a filtration plate of this kind can beused simultaneously as a receiver and as a packaging material for theplutonium oxide deposit, so that the latter can be transported, withoutsubsequent treatment, to the place where the final purification iscarried out.

The method according to the invention will be more fully explained bythe following examples:

EXAMPLE 1 Sintered tablets of U0 and 15 percent by weight of PuOobtained by coprecipitation and sintering, and

weighing approximately 1.8 g each were used as starting material. Atable of this type was placed in a molten salt bath (approximately 4 g)consisting of a eutectic mixture of NaNO;, and KNO at 450 C. After areaction period of 3 hours, the tablet was completely disintegrated anda brownish pulverulent deposit had formed. 58 g of a eutectic mixture ofKCl and LiCl were added at the same temperature and a mixture of 20percent by volume of HCl and 80 percent by volume of was passed throughfor a period of 2% hours. After a decantation period of IO minutes, asample of about 3 g was taken from the melt. Analysis showed that noplutonium could be detected in the melt, while no uranium was present inthe deposit.

EXAMPLE 2 This example was carried out under similar conditions to thosedescribed in Example 1 except that 1.32 g tablets were used consistingof U0; and 20 percent by weight of PuO obtained by mechanically mixingthe oxide powders and sintering.

No plutonium could be detected in the melt after sampling, while theuranium had completely dissolved.

We claim:

